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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20141K9961997-05-0505 May 1997 Safety Evaluation Supporting Amend 89 to License DPR-34 ML20128D7191992-12-0101 December 1992 Safety Evaluation Approving Exemption from Requirement of 10CFR50.54(q) to Change to Biennial Emergency Plan Exercise Rather than Annual Following Completion of Next Scheduled Exercise at Plant ML20246J3261989-08-30030 August 1989 Safety Evaluation Supporting Amend 72 to License DPR-34 ML20245J3781989-08-14014 August 1989 Safety Evaluation Supporting Amend 71 to License DPR-34 ML20245J4511989-08-0808 August 1989 Safety Evaluation Responding to Issues Re Tech Spec Upgrade & Plant Defueling.Stated Tech Spec Sections Should Be Upgraded ML20246J3131989-07-0707 July 1989 Safety Evaluation Concluding That Operators Role in Mitigating High Energy Line Break at Facility Acceptable ML20247R2261989-05-26026 May 1989 Final Safety Evaluation Re LER 87-20 Concerning Interactions Between Steamline Rupture Detection/Isolation Sys,Plant Protective Sys & Control Sys at Facility ML20245C5031989-04-18018 April 1989 Safety Evaluation Supporting Amend 70 to License DPR-34 ML20248D6501989-03-31031 March 1989 Safety Evaluation Supporting Amend 69 to License DPR-34 ML20236A1401989-02-27027 February 1989 Safety Evaluation Supporting Amend 68 to License DPR-34 ML20235T4511989-02-24024 February 1989 Safety Evaluation Re Facility Core Support Flow Vent Sys. Continued Operation of Facility W/Current Core Support Flow Sys Configuration Acceptable ML20235J3421989-02-16016 February 1989 Safety Evaluation Supporting Util Action in Response to Generic Ltr 83-28,item 2.1 (Part 2) Confirming Establishment of Interface W/Either NSSS Vendor or Vendors of Each Component in Reactor Trip Sys ML20235J3841989-02-13013 February 1989 Safety Evaluation Concluding That Continued Operation of Facility Not Affected by Steam Generator Tube Failures Experienced by Advanced gas-cooled Reactors ML20195D3911988-10-27027 October 1988 Safety Evaluation Supporting Corrective Actions of LER 86-017 ML20205G0021988-10-24024 October 1988 Safety Evaluation Supporting Amend 65 to License DPR-34 ML20154J8021988-09-15015 September 1988 Safety Evaluation Supporting Amend 64 to License DPR-34 ML20154J4621988-09-15015 September 1988 Safety Evaluation Supporting Amend 63 to License DPR-34 ML20207F0571988-08-10010 August 1988 Safety Evaluation Supporting Util 870206 Submittal Re Safe Shutdowns During Postulated Accident Conditions ML20207F0431988-08-0505 August 1988 Safety Evaluation Supporting Amend 61 to License DPR-34 ML20207F2411988-08-0505 August 1988 Safety Evaluation Supporting Amend 62 to License DPR-34 ML20151M1601988-07-21021 July 1988 Safety Evaluating Supporting Requirements for Redundancy in Responding to Rapid Depressurization Accident ML20151A9961988-06-20020 June 1988 Safety Evaluation Supporting Amend 60 to License DPR-34 ML20195K0651988-06-15015 June 1988 SER Concurring W/Util Proposed Corrective Actions in Engineering Rept Entitled, Rept of Helium Circulator S/N 2101 Damage & Inlet Piping S/N 2001 Repair & Mod Activities ML20195F9661988-06-15015 June 1988 Safety Evaluation Re Destructive Exam Rept for Fuel Test Assembly-2.Fuel Represented by Fuel Test Assembly-2 Predicted to Be Safe for Operation in Facility for 1,800 EFPDs ML20154F8891988-05-10010 May 1988 Safety Evaluation Re Proposed Safe Shutdown Sys & Exemption Requests Concerning 10CFR50,App R.Licensee Request for Exemptions in Listed Areas Should Be Granted.Concept for Providing post-fire Shutdown Acceptable ML20148S6031988-04-0707 April 1988 Safety Evaluation Supporting Amend 59 to License DPR-34 ML20151B6651988-04-0101 April 1988 Supplemental Safety Evaluation Supporting Util Compliance w/10CFR50.App R Re Safe Shutdown DHR Capacity ML20150C4541988-03-10010 March 1988 Safety Evaluation Concluding That Seismic Analysis Methods for Bldg 10 & Walkover Structure Conservative.Gaps Provided Adequate to Accommodate Relative Motions Which Occur Between Subj Structures & Walkover Structure & Turbine Bldg ML20147C8181988-02-25025 February 1988 Safety Evaluation Supporting Changes to Interim Tech Specs 3/4.1.7, Reactivity Change W/Temp NUREG-1220, Safety Evaluation Accepting Plant Special Senior Licensed Fuel Handler Initial & Requalification Operator Training Program,Per NUREG-1220, Training Review Criteria & Procedures1988-01-13013 January 1988 Safety Evaluation Accepting Plant Special Senior Licensed Fuel Handler Initial & Requalification Operator Training Program,Per NUREG-1220, Training Review Criteria & Procedures ML20237D7631987-12-18018 December 1987 Safety Evaluation Updating 861118 Fire Protection Sys Safety Evaluation.Util Alternate Fire Protection Configuration Acceptable ML20149E1621987-12-18018 December 1987 Marked-up Safety Evaluation Re Proposed Safe Shutdown Sys & Exemption Requests Concerning 10CFR50,App R ML20236U6961987-11-23023 November 1987 Safety Evaluation Supporting Amend 57 to License DPR-34. Addendum to Review of Proposed Tech Spec Change:Core Inlet Valves/Min Helium Flow & Max Core Region Temp Rise,Limiting Condition for Operation..., Technical Evaluation Rept Encl ML20236U5761987-11-20020 November 1987 Safety Evaluation Re Helium Circulator S/N C-2101 Damage. Util Corrective Action Program Initiated ML20236R3001987-11-13013 November 1987 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 2.2.1 Re Equipment Classification Programs for All safety-related Components ML20238C7621987-09-0202 September 1987 Safety Evaluation Concurring W/Util 870702 & 27 Ltrs & 870818 Telcon Re Elimination or Reduction of Maint Requirements on Certain Fire Seals ML20235N6491987-07-13013 July 1987 Safety Evaluation Supporting Amend 56 to License DPR-34 ML20235F5281987-07-0202 July 1987 Safety Evaluation Re Safe Shutdown of Steam Generator Matls. Under Severe Transient Conditions,Fuel Temp Can Be Maintained Under Accepted Temp Limits & Plant Can Be Safely Shutdown ML20235F5151987-07-0202 July 1987 Safety Evaluation Re Safe Emergency Shutdown of Reactor Sys. Operation at 82% Acceptable ML20235F5441987-07-0202 July 1987 Safety Evaluation Re Effect of Firewater Cooldown on Steam Generator Structural Integrity.All Tests Acceptable ML20235E5281987-06-29029 June 1987 Safety Evaluation Supporting Amend 55 to License DPR-34 ML20216G9511987-06-24024 June 1987 Revised Safety Evaluation Re Steam Line Rupture Detection & Isolation Sys (Slrdis).Slrdis Meets Requirements of 10CFR50, App A,Gdc 20 & GDC 4 ML20216G9911987-06-24024 June 1987 Supplemental Safety Evaluation Supporting Application for Amend to License DPR-34 Re Tech Specs for Steam Line Rupture Detection & Isolation Sys ML20215J5401987-06-22022 June 1987 Draft Safety Evaluation Re Safe Emergency Shutdowns.Facility Operation at 82% Power Acceptable ML20216J1921987-06-17017 June 1987 Safety Evaluation Re Mods to Reduce Moisture Ingress Into Reactor Vessel.Periodic Insps & Preventive Maint Should Be Performed on Pertinent Components.Operational Performance Should Be Continuously Upgraded ML20214M4681987-05-20020 May 1987 Safety Evaluation Supporting Amend 54 to License DPR-34 ML20215J8271987-05-0505 May 1987 Safety Evaluation Supporting Amend 53 to License DPR-34 ML20209D7561987-04-22022 April 1987 Safety Evaluation Supporting Util 870211 Submittal Re Performance Enhancement Program,Finding 4-10 ML20206J9331987-04-0606 April 1987 Safety Evaluation Supporting Amend 52 to License DPR-34 ML20205S1141987-03-31031 March 1987 Safety Evaluation Accepting Util 831104 Response to Generic Ltr 83-28,Item 4.5.2, Reactor Trip Sys Reliability On-Line Testing. Facility Designed to Permit on-line Functional Testing,Including Testing of Reactor Trip Contactors 1997-05-05
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20196G6731997-07-0101 July 1997 Informs Commission That Decommissioning Process Has Been Completed at PSC of Colorado Fsvngs,Unit 1 Located in Town of Platteville in Weld County,Co ML20141K9961997-05-0505 May 1997 Safety Evaluation Supporting Amend 89 to License DPR-34 ML20140E1121997-04-10010 April 1997 Confirmatory Survey of Group Effluent Discharge Pathway Areas for Fsv Nuclear Station,Platteville,Co ML20134D1661997-01-30030 January 1997 Rev 1,Vol 6 to Final Survey Rept,Final Survey of Group E (Book 2A of 2) ML20137S6111996-12-31031 December 1996 Annual Rept Pursuant to Section 13 or 15(d) of Securities Exchange Act 1934, for Fy Ended Dec 1996 ML20134G6401996-10-29029 October 1996 Rev 0,Volume 6,Books 1 & 2 of 2 to Final Survey of Group E ML20134G6171996-10-29029 October 1996 Rev 2,Volume 1,Books 1 & 2 of 2 to Final Survey Description & Results ML20134G7271996-10-29029 October 1996 Rev 0,Volume 11,Book 1 of 1 to Final Survey of Group J ML20134G6861996-10-29029 October 1996 Rev 0,Volume 8,Books 1 & 2 of 2 to, Final Survey of Group G ML20134G6321996-10-26026 October 1996 Rev 1,Volume 5,Books 2 & 3 of 3 to Final Survey of Group D ML20133D7831996-10-22022 October 1996 Preliminary Rept - Orise Support of NRC License Insp at Fsv on 960930-1003 ML20116A4661996-07-19019 July 1996 Fsv Final Survey Exposure Rate Measurements ML20112J6861996-05-31031 May 1996 June 1996 Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments for Fsv Decommissioning.Rept Covers Period of 960216-0531 ML20112C1531996-05-17017 May 1996 Fsv Final Survey Exposure Rate Measurements ML20101G5521996-03-21021 March 1996 Confirmatory Survey Activities for Fsv Nuclear Station PSC Platteville,Co, Final Rept ML20097E3201996-01-31031 January 1996 Nonproprietary Fort St Vrain Technical Basis Documents for Piping Survey Instrumentation ML20095K4131995-12-26026 December 1995 Rev 3 to Decommissioning Plan ML20095H7211995-12-20020 December 1995 Revs to Fort St Vrain Decommissioning Fire Protection Plan Update ML20095K9751995-12-15015 December 1995 Fort St Vrain Project Update Presentation to NRC, on 951207 & 15 ML20096C1671995-12-13013 December 1995 Rev 4 to Decommissioning Fire Protection Plan ML20094M1651995-11-30030 November 1995 Nonproprietary Fsv Technical Basis Documents for Piping Survey Implementation ML20092F3461995-09-14014 September 1995 Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments for Fsv Decommissioning, Covering Period of 950516-0815.W/ ML20137H3531994-12-31031 December 1994 Partially Withheld, Rept of Independent Counsel Investigation Concerning Issues at Fort St Vrain Nuclear Generating Station Decommissioning Project, App D,Comments by Mkf & Westinghouse Team & Responses ML20137S2331994-12-31031 December 1994 Rept of Independent Counsel Investigation Concerning Issues at Fort St Vrain Nuclear Generating Station Decommissioning Project, Dec 1994 ML20029C6031993-12-31031 December 1993 1993 Annual Rept Public Svc Co of Colorado. W/940405 Ltr ML20058Q3791993-12-21021 December 1993 Rev 1 to Decommissioning Plan for Fort St Vrain Nuclear Generating Station ML20045B3641993-06-30030 June 1993 June 1993 Quarterly 10CFR50.59 Rept of Changes,Tests & Experiments for Fsv Decommissioning. ML20045A4291993-06-0303 June 1993 LER 93-003-00:on 930505,new Source of Natural Gas Introduced within 0.5 Miles of ISFSI & Reactor Bldg W/O Prior NRC Approval.Caused by Field Routing of Natural Gas Pipe.Well Isolated by Well operator.W/930603 Ltr ML20077D1631993-05-10010 May 1993 Enforcement Conference, in Arlington,Tx ML20125C7161992-12-0707 December 1992 Part 21 Rept Re Possibility for Malfunction of Declutching Mechanisms in SMB/SB-000 & SMB/SB/SBD-00 Actuators. Malfunction Only Occurs During Seismic Event.Balanced Levers May Be Purchased from Vendor.List of Affected Utils Encl ML20128D7191992-12-0101 December 1992 Safety Evaluation Approving Exemption from Requirement of 10CFR50.54(q) to Change to Biennial Emergency Plan Exercise Rather than Annual Following Completion of Next Scheduled Exercise at Plant ML20127P5861992-11-23023 November 1992 Followup to 921005 Part 21 Rept Re Potential Defect in SB/SBD-1 Housing Cover Screws.Procedure Re Replacement of SBD-1 Spring Cover Bolts Encl.All Fasteners Should Be Loosened & Removed.List of Affected Utils Encl ML20127F5691992-11-0303 November 1992 Informs Commission of Intent to Issue Order Approving Plant Decommissioning Plan & Corresponding Amend to License DPR-34 ML20101E5761992-05-31031 May 1992 Monthly Defueling Operations Rept for May 1992 for Fort St Vrain ML20096E8221992-04-30030 April 1992 Monthly Operating Rept for Apr 1992 for Fort St Vrain.W/ ML20095E9601992-04-17017 April 1992 Rev to Fort St Vrain Proposed Decommissioning Plan ML20100R7431992-03-31031 March 1992 Monthly Operating Rept for Mar 1992 for Fort St Vrain.W/ ML20090L0621992-02-29029 February 1992 Monthly Operating Rept for Feb 1992 for Fort St Vrain Unit 1 ML20092D0081992-01-31031 January 1992 Monthly Operating Rept for Jan 1992 for Fort St Vrain Nuclear Generating Station ML20102B2241992-01-22022 January 1992 Fort St Vrain Station Annual Rept of Changes,Tests & Experiments Not Requiring Prior Commission Approval Per 10CFR50.59, for Period 910123-920122 ML20094N6701991-12-31031 December 1991 Public Svc Co Annual Financial Rept for 1991 ML20091J6251991-12-31031 December 1991 Monthly Operating Rept for Dec 1991 for Fort St Vrain.W/ ML20094D6711991-11-30030 November 1991 Monthly Operating Rept for Nov 1991 for Fort St Vrain Unit 1 ML20090M1871991-11-20020 November 1991 FOSAVEX-91 Scenario for 1991 Plant Exercise of Defueling Emergency Response Plan ML20086D6891991-11-15015 November 1991 Proposed Decommissioning Plan for Fort St Vrain Nuclear Generating Station ML20085N1451991-11-0505 November 1991 Revised Ro:Operability Date of 910830 for Electric Motor Driven Fire Water Pump P-4501 Not Met.Pump Not Actually Declared Operable Until 911025.Caused by Unforseen Matl & Testing Problems.Equivalent Pump Available ML20086C5451991-10-31031 October 1991 Monthly Operating Rept for Oct 1991 for Fort St Vrain.W/ ML20085H6611991-10-10010 October 1991 Assessment of Mgt Modes for Graphite from Reactor Decommissioning ML20091D7671991-10-0101 October 1991 Rev B to Engineering Evaluation of Prestressed Concrete Reactor Vessel & Core Support Floor Structures for Proposed Sys 46 Temp Change ML20085D9861991-09-30030 September 1991 Monthly Operating Rept for Sept 1991 for Fort St Vrain.W/ 1997-07-01
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' FINAL = SAFETY EVALUATION BY :THE OFFICE OF: NUCLEAR REACTOR REGULATION RELATED TC LER 87-20
' INTERACTIONS OF FORT ST. VRAIN PFS/SLRDIS WITH CONTROL SYSTEMS FORT ST. VRAIN NUCLEAR GENERATING STATION PUBLIC SERVICE COMPANY OF COLORADO U " DGCKET N0; 50-267 1.0; INTPODUCTI0f? Afl0 BACKGROUND The Fcrt St. Vrain (FSV) Plant has originally designed with a Plant.Protec- >
tive System (PPS) providing various levels of. protection to the . reactor and specific equipment iters against abnornial crerating ccrditions. One featurc of
.the plant prctective system was the 1 cop shutdewn, which automatically isolated one of the two secondary coolant loops under certain conditions. The PPS automat-ically closed a number of valves in. e loop as part'of the loop shutdown.
'In 1986, the licensee added a second autcratic prctection syster, the . . .
Steam.Line Rupture Petection/ Isolation System (SLECIS). The SLRDIS is designed to'pretect qualified equipment in the plant from the. harsh environment that-could be created by an urritigated high energy lire break. SLRDIS senses am- .
bient temperatures in the Reactor and Turbine Building and then icclates high energy-lines in both loops' by closing a number of valves. SLRDIS utilizes '
certain tristing PPS circuitry'for this purpose, as well as new circuitry added as part of the SLRDIS redifications. The staff had previously reviewed and .
-approved the'SLRDIS in a safety evaluation (SE) dated February 26,19E7. >
Approval of SLRDIS was also discussed in a SE dated April 3,1987 on the -
licensee's equipment qualification program.
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However, in a licensee event report (LER) dated October 1,1987 (LER 87-020) '
the licensee stated:
"First-in-with-lockout logic is provided into the PPS to prevent autcratic shutdown of both-loops. This lockout feature is overridden in the event of a high energy line break by SLRDIS, which automatically closes the feedwater block and control valves of both loops simultaneously. There are also interlocks provided in the control circuits of the feedwater .
. block valves and feedwater ficw control valves which prevent closing both loop feedwater control valves at the sare time."
l f .In the requests for additional information dated September 20, 1988 and January 12, 1989, the staff indicated their corcern about the potential interacticrs cf the PPS and control system, both of which provide outputs to the feedwater b1cck and flow control valves.
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In particular, the staff was concerned that the control circuits could override FPS or SLRDIS acticos aric prevent accomplishment of the required
- protective functions, ar.d requested a review of these potential interactier,s against Criterion 19, 20, 21, and 22 of the FSV Final Safety Analysis Report (FSAR).
By letters dated November 22, 1988 ano April 7, 1989, the licensee responded to these concerns.' The following evaluation covers this response.
2.0 EVALUATION The licensee was requested to review the interaction between the SLRDIS and PPS systems and the control system, that could prevent the accomplishment'of the required protective functions. The licensee's evaluation and this evaluation are done against criterion 19, 20, 21 and 22 of the FSV FSAR.
2.1 Criterion 19 The licensee has stated that the SLRDIS system is built to be highly reliable and testable.
The staff is in general acreement with these conclusicos. However, the staff hadnotedinSectionV.2.Ic)ofthelicensee'sreportofNovember 22, 1988, that the SLRDIS relay contacts provide a zero input to the velve position card to close the control valve. However, the provisions to assure that other inputs to the control card er failure of the control-card power supplies, or the centrol card itself were not addressed. This problem could potentially exist for the following groups of valves:
Circulator speed controls, Feedwater flow controls, and Reheat steam bypass pressure ratio control.
The most significant concern was the possibility of multiple failures resulting from failures of a commen power supply. By letter dated April 7,1989, the licensee submitted a more detailed arialysis of the failure modes for these valves. Operation of these valves is based on a combination of electrical, hydraulic and instrument (compressed) air power sources. Both the hydraulic and instrument air systems are designed (and operated) to be single failure proof. Therefore, failures of these power sources did not pose a problem for the SLRDIS system. Power sources for the valves control cards and the electrical control solencids are diverse, with supplies from both instrument bases.
Failure of an individual power supply component (i.e., fuse), or an essential bus can prevent an individual SLRDIS valves from closing. However, by the analysis presented in the licensee's April 7,1989 submittal, the licensee had demonstrated that another valve (s) in the system will close and provide the j needed isolation function.
l In view of the above, the staff concludes that the desired SLRDIS function I
will be performed with reasonable assurance. Thus, the staff concludes the SLRDIS system meets criterion 19 of the FSV FSAR.
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l 2.2 Criterion 20 i
The licensee has stated that the SLRDIS system is sufficiently redundant and independent such that no single failure will prevent an overall system failure or lead to multiple SLRDIS valves failing to close. As noted in the discussion for Criterion 19, the SLRDIS input to the valve positicn cards for certain valves have been analyzed fer common failure modes. Consideration has also ;
been given to failure rrodes of the various power supplies. Based on the licensee's analysis, the staff concludes the SLRDIS systeni n'eets Criterion 20.
2.3 Criterion 21 The licensee has stated that the objective of the SLRDIS system is met for the case of a single valve or corpenent failure since the valves actuated by SLRDIS are actuated by both SLRDIS loops and are in series with another valve actuated by the other SLRDIS loop. The staff agrees in general that the 4 SLRDIS design is redundant in that multiple valves are closed in each loop, and that this limits the total volume (mass) of high energy fluid that ten escape thrcugh any single line break.
Where valves fail to close due to failure of a power supply, redundant adjacent ,
valves has been identified to perforrn the required function. Thus, there is no pctential for multiple valve failures and the design cbjective is met. Therefore, the staff concludes Criterion 21 is satisfied.
2.4 Criterion 22 The licensee has stated that by the use of isolated centacts, no failure cf I the control circuit can propagate back into the PPS or SLRDIS systems. The staff agrees with this conclusion. The staff has also examined the use of the
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isolated contacts to control the position of the open or shut type control circuits as indicated in the licensee's discussion of HV-2201, the Loop Feedwater Block Valve. In this circuit, the isolated SLRDIS contacts provide an overriding signal for the closure of the subject valve. Therefore, the staff concludes that Criterion 22 is satisfied.
2.5 Other Considerations i The staff notes that System Description-93-2, entitled "Overall Plant Control I and Plant Protective System" states that on the trip of both circulators in a '
loop results in closure of the reheater inlet attemporator line block and flow j control valves. It is the staff's understanding that the attemporator inlet !
lines are small, and therefore the closure of these valves by SLRDIS is net !
required to mitigate the effects of a line break. The licensee confirmed the role of these valves in the operation cf the SLRDIS system in the April 7, 1989 submittal. ,
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3.0 CONCLUSION
The staff concludes that the SLRDIS design provides adequate isolation between l the PPS and control system functions. The staff has reasonable assurance feilures 1 of the control syster will not affect PPS operation. Furtherraore, the SLRDIS l
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system will properly actuate _ the desired valves. to close and. reduce the effects -
. of a high-energy -line break. In cases where a' power supply failure'does not -
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" ~ allcw a SLRDIS valve'to close', a ' redundant valve in.close proximity is closed and the necessary isolation provided.
Dated: May 26, 1.989
' Principal' Contributor: Kenneth L. Heitner, PD-IV-
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